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Advanced Power Reactor of 1400 MWe (APR1400) adopts a Direct Vessel Injection (DVI) system that injects Safety Injection (SI) water directly into the reactor vessel downcomer during accidents. As the DVI nozzles are directly attached to the reactor vessel downcomer, complex thermal-hydraulic phenomena occur in the downcomer region. Recently, in the International Standard Problem (ISP) No. 50 exercise, the phenomena of Emergency Core Cooling (ECC) water mixing in the upper downcomer during the DVI line break accident observed in Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) test were highlighted in terms of the prediction capability of the system analysis codes. Thus, to validate the results from the ATLAS study independently, an additional experimental study was performed with an integral effect test facility, Seoul National University Facility (SNUF), to observe the mixing behavior in the downcomer during a DVI line break accident. According to the SNUF test results, the ECC water mixed vigorously in the downcomer annulus. However, the temperature difference in the azimuthal direction was predicted by a system analysis code, Multi-dimensional Analysis Reactor Safety (MARS). In the MARS calculation, the momentum flux terms are set to zero for the junction between the one-dimensional volume and three-dimensional cell of the MultiD component because the axial and radial velocities are marginal in the large three-dimensional region. However, if the nozzles are attached to the downcomer with a thin gap, the axial and radial velocities are significant when the incoming orthogonal flow through the nozzles impinges against the downcomer wall. It was necessary to consider the momentum flux terms induced by the impinging flow, and to do so, an appropriate jet impingement model, for incorporation in the system analysis code, MARS, was developed in this study. To develop the jet impingement model, Computational Fluid Dynamics (CFD) calculations were carried out, and the jet impingement model was formulated based on the CFD calculations under various conditions. The momentum flux term resulting from the jet impingement phenomenon was correlated with the diameter of the nozzle, downcomer gap size, and incoming flow velocity. This model was applied to MARS by considering the momentum flux term for the junctions connected to the cell of the MultiD component. The modified MARS incorporating the jet impingement model was validated with the test results from the SNUF and ATLAS, and the analysis results exhibited reasonable agreement with the test data.

In nuclear power plants (NPPs), one of the important yet uncertain phenomena during a station blackout (SBO) accident is the possibility for a counter-current natural circulation flow in the reactor coolant system (RCS) pressure boundary, which could divert the core melt progression and/or the in-vessel fission product (FP) distribution and, in turn, lead to a heatup and eventual creep rupture. From the risk and accident management points of view, one of key concerns is to predict more explicitly whether such a creep rupture can occur in the RCS pressure boundary, and if it indeed can, also predict the most probable time and the most likely location. This paper presents plant-specific analysis results on the potential characteristics of natural circulation-induced creep rupture of the RCS pressure boundary that could potentially take place during an SBO accident of a reference pressurized water reactor (APR1400). MELCOR-CFD (Computational Fluid Dynamics) coupled analysis was performed to determine more precisely the relevant physical parameter that could greatly influence on the behaviour of natural circulation and the ensuing creep rupture. Multiple sensitivity analyses were also carried out to investigate the influence that particular MELCOR modeling schemes and plant conditions have in estimating the RCS creep rupture characteristics. The analysis results and relevant insights were summarized in terms of particular points of interest.

A steam generator tube rupture (SGTR) is a type of containment bypass accidents that if core damage is not prevented during this accident progression, has the potential to lead to a significant release of fission products (FPs) to the environment through the faulted steam generator (SG) or other possible paths in the secondary system. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during an SGTR accident of a reference pressurized water reactor (PWR) plant, and also investigates the impact of dedicated accident mitigation actions on the results of interest through the established modeling approach. To consider the integral responses of the reactor, SGs, and containment to a severe accident initiated by SGTR, detailed modeling was performed for a broad spectrum of plant safety/mitigation systems and relevant phenomena influencing the evolution of the accident. As the modeling and simulation tools for the SGTR accident analysis, MELCOR 2.2 and MAAP5 were employed to compare the results of interest. The results of the analysis demonstrate that the release of FPs into the environment could be estimated more realistically as a result of the best practice modeling for the SGTR accident and also it could be greatly decreased through dedicated mitigation actions. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

An interfacing systems loss of coolant accident (ISLOCA) is a type of containment bypass accident that may lead to a significant release of fission products (FPs) to the environment. This paper presents a best practice modeling approach to more realistically quantify the FP source terms expected during the ISLOCA of a reference pressurized water reactor (PWR) plant and investigates the influence of a dedicated accident mitigation action on the results of interest through the established modeling approach. In addition to this, the key plant features and phenomenological mechanisms that might have a significant influence on the magnitude and timing of FP releases to the environment during an ISLOCA were examined through a couple of sensitivity analyses. To consider the integral responses of the reactor, containment, and auxiliary building to a severe accident initiated by an ISLOCA, detailed modeling was performed for a broad spectrum of relevant phenomena, plant systems, and structures that promote the deposition and retention of FPs along the flow path to the environment. As the modeling and simulation tools for the ISLOCA analysis, MELCOR 2.2 and MAAP5.04 were employed to compare the results of interest. The results of the analysis shows that the present best practice modeling approach proposed here has the potential to significantly decrease the release of FPs into the environment in the event of an ISLOCA. Relevant insights are summarized in terms of particular points of interest and remaining uncertainty issues for further study.

KAERI has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), for accident simulations of advanced PWRs. Regarding integral effect tests, a database for major design basis accidents has been accumulated and a Domestic Standard Problem (DSP) exercise using the ATLAS has been proposed and successfully performed. The ATLAS DSP aims at the effective utilization of an integral effect database obtained from the ATLAS, the establishment of a cooperative framework in the domestic nuclear industry, better understanding of thermal hydraulic phenomena, and an investigation of the potential limitations of the existing best-estimate safety analysis codes. For the first ATLAS DSP exercise (DSP-01), integral effect test data for a 100% DVI line break accident of the APR1400 was selected by considering its technical importance and by incorporating comments from participants. Twelve domestic organizations joined in this DSP-01 exercise. Finally, ten of these organizations submitted their calculation results. This ATLAS DSP-01 exercise progressed as an open calculation; the integral effect test data was delivered to the participants prior to the code calculations. The MARS-KS was favored by most participants but the RELAP5/MOD3.3 code was also used by a few participants. This paper presents all the information of the DSP-01 exercise as well as the comparison results between the calculations and the test data. Lessons learned from the first DSP-01 are presented and recommendations for code users as well as for developers are suggested.

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

The transient thermal–hydraulic phenomena of a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in pressurized water reactor, APR1400, were investigated. In order to understand the phenomena during the LOCA transient, a reduced-height and reduced-pressure integral loop test facility, the SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment with the SNUF, the energy scaling method was suggested with scaling the coolant mass inventory and the thermal power for the reduced-pressure condition. According to the conditions determined by the method, the experimental study was performed with the SNUF. The experimental results showed that the phenomenon of the downcomer seal clearing played a dominant role in the reduction of the system pressure and the recovery of the coolant level in the core. That phenomenon occurred when the steam incoming from cold legs penetrates the coolant in the upper downcomer toward the broken DVI line. The experimental results were compared with the prototype analysis to estimate the energy scaling method, so that the experiment reasonably simulated the phenomena in the prototype. For the analytical investigation, the experiment was simulated with MARS code to validate the calculation capability of the code, especially for the downcomer seal clearing, which showed good agreement with the results of experiment.

A scaling methodology for a small-scale integral test facility was investigated in order to analyze thermal–hydraulic phenomena during a DVI (direct vessel injection) line SBLOCA (small break loss-of-coolant accident) in an APR1400 (advanced power reactor 1400 MWe) pressurized water reactor. The test facility SNUF (Seoul National University Facility) was utilized as a reduced-height and reduced-pressure integral test loop. To determine suitable test conditions for simulating the prototype in the SNUF experiment, the energy scaling methodology was propose to scale the coolant mass inventory and the thermal power for a reduced-pressure condition. The energy scaling methodology was validated with a system code (MARS) analysis for an ideally scaled-down SNUF model and that predicted a reasonable transient of pressure and coolant inventory when compared to the prototype model. For the actually constructed SNUF, the effect of scaling distortions in the test facility's thermal power and the loop geometry was analytically investigated. To overcome the limitation of the thermal power supply in the facility, the convective heat transfer between primary and secondary systems at the steam generator U-tubes was excluded and a modified power curve was applied for simulating the core decay heat. From the code analysis results for the actual SNUF model, the application of the modified power curve did not affect the major events occurring during the transient condition. The results revealed that the scaling distortion in the actual SNUF geometry also did not strongly disturb significant thermal–hydraulic phenomena such as the downcomer seal clearing. Thus, with an adoption of the energy scaling methodology, the thermal–hydraulic phenomena observed in the SNUF experiment can be properly utilized in a safety analysis for a DVI line break SBLOCA in the APR1400.